Subcritical Calculations Capabilities

User Support
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Hello everyone,

I’m new to OpenMC and currently working on my very first project using the software. I’m quite curious about the various capabilities it offers. Specifically, I’m interested in adding an intrinsic neutron source with spontaneous fission, (a-n) fission, and spallation neutron source capabilities. I’m wondering if OpenMC has the capability to model such a neutron source.

Additionally, I’m keen to perform an analysis on a subcritical reactor. I would like to explore the capabilities related to Subcritical Power Calculations and the Subcritical Multiplication Factor (ks). I’m curious about the accuracy of modeling these aspects and how I can extract the necessary data for comparison with the normal running output of the k-eff. My research aims to understand any divergence between these values in a subcritical reactor.

Any help would be highly appreciated. Thank you in advance!

2

Hi @jimmyvdw and welcome to the community! OpenMC does not currently have any ability to model spontaneous fission, (alpha,n), or spallation neutron sources. You would need to generate a neutron source description (spatial, energy, and angle distributions of neutrons) from another code and then feed that into OpenMC if you wanted to study the neutron behavior.

Subcritical multiplication calculations are possible by running in fixed source mode with fissionable materials in your model.

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Hello @paulromano, thank you for the clarification. When it comes to modelling the neutron source description, how would I integrate this file into the code (say a file generated with Geant4) and what type of input file is taken in OpenMC.